ML14183A781
| ML14183A781 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 01/15/1997 |
| From: | Mozafari B NRC (Affiliation Not Assigned) |
| To: | Hinnant C CAROLINA POWER & LIGHT CO. |
| References | |
| TAC-M96355, NUDOCS 9701230264 | |
| Download: ML14183A781 (6) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 15, 1997 Mr. C. S. Hinnant, Vice President Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Unit No. 2 3581 West Entrance Road Hartsville, South Carolina 29550
SUBJECT:
10 CFR 50.46 LARGE BREAK LOSS OF COOLANT EVALUATION MODEL FOR H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 (TAC M96355)
Dear Mr. Hinnant:
In our letter of October 11, 1996, we informed you about the problems that we had identified concerning changes in the large break loss-of-coolant accident (LBLOCA) evaluation model developed by Siemens Power Corporation (SPC) for pressurized water reactors (PWRs) to comply with 10 CFR 50.46. The changes to the 1986 approved LBLOCA evaluation model were proposed to correct an error of nonphysical behavior in the prediction of heat transfer coefficients during core reflood. The range of concern for core reflood rates following the LBLOCA is for rates of 1.0 in/sec to 1.77 in/sec. Within this range the 1986 model predicted a peak in the heat transfer coefficients and then the nonphysical behavior of decreasing coefficients with increasing reflood rates.
The SPC 1991 model, which was also discussed in the October 11, 1996, letter, is not an acceptable LBLOCA model.
We stated in the letter of October 11, 1996, that the 1986 model had an unacceptable error and we requested, in accordance with 10 CFR 50.46(a)(3)(ii), that you assess the impact of the model error and changes, and take whatever actions are required to assure compliance with 10 CFR 50.46 for the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBR).
You were also requested to attend a public meeting on October 16, 1996, at the Nuclear Regulatory Commission headquarters to present the results of the assessment of the peak cladding temperature (PCT) for the LBLOCA and the corrective actions and compensatory measures that have been undertaken, both short-term and long term to demonstrate compliance with 10 CFR 50.46. An acceptable short-term measure would be to use a demonstrably conservative reflood heat transfer coefficient model; an acceptable long-term measure must be to eliminate the nonphysical behavior from the model.
In both cases, the model must comply with Appendix K to 10 CFR Part 50. The summary of the October 16, 1996, meeting was issued on November 5, 1996.
You submitted letters to address this issue on October 14, 25, and 29, 1996.
In the letter of October 14, 1996, you submitted notification of a significant error (i.e., resulting in a greater than 50 OF change in calculated PCT) in the LBLOCA model for HBR. The letter addressed the impact of the 1991 model not being acceptable and, because the reflood rates for HBR are within the range of concern, the nonphysical behavior of the 1986 model. The 1986 model was modified to establish the predicted heat transfer coefficients for the 2 30 0 25 a
CETRCOPY 9701230264 970115 PDR ADOCK 05000261 P
Mr. C. S. Hinnant
-2 reflood range of 0 in/sec to 1.77 in/sec at no greater than the values calculated for the 1.77 in/sec reflood rate. Based on an estimate of the PCT, you stated that compensatory measures were incorporated into the core operating limits report (COLR) to reduce the allowable heat flux hot channel factor and nuclear enthalpy rise hot channel factor. The new estimated PCT based on the compensatory measures is 2157 oF, an increase of 93 OF.
You further stated that this estimate is valid for approximately 87 effective full power days (EFPDs) of operation.
The letter of October 25, 1996, documented what you had presented at the October 16, 1996, meeting regarding your assessment of the PCT for the LBLOCA and the corrective actions and compensatory measures that have been undertaken, both short-term and long-term for HBR. In that letter, you stated that the information presented in the letter of October 14, 1996, was also presented in the meeting of October 16, 1996, and you referenced the conference call held on October 23, 1996, during which SPC provided clarification about its discussion of conservatism in the 1986 model in the October 16, 1996, meeting. In the meeting, the discussion was on the use of a "modified" 1986 model that capped the heat transfer coefficients at the coefficient value for 1.77 in/sec for the reflood range of concern to correct the error. Based on the new data shown in the meeting by SPC about the significant conservatism in the modified 1986 model, the NRC staff concluded in that meeting that the licensees involved had taken those actions required by 10 CFR 50.46 to allow continued plant operation. We, however, requested that the licensees at the meeting submit all the data that show the modified 1986 model is conservative over the entire range of reflood rates of concern so that the staff could review the data to determine an acceptable correction to the error in the 1986 model.
In the conference call on October 23, 1996, we were informed by SPC that the 1986 model was not as conservative with respect to the measured heat transfer coefficients as was presented in the October 16 meeting. Based on this new information, the staff has concluded that "capping" the reflood heat transfer coefficient in the range of concern at the value at a 1.77 in/sec reflood rate has not been demonstrated to be conservative. The staff has informed SPC that the nonphysical error in the 1986 model must be corrected using a method that is demonstrably conservative. SPC submitted this correction on December 20, 1996, and the NRC staff is reviewing it.
As stated in the letter of October 25, 1996, you reassessed the nonphysical behavior of the 1986 model and replaced this model with the heat transfer coefficients correlation from the FLECHT SEASET data over the entire range of reflood rates of concern, 0 in/sec to 1.77 in/sec. The estimated PCT is 2128 OF. This is based on the compensatory measures listed in the letter of October 14, 1996, and limiting core power to 76 percent rated thermal power when using the 5 percent axial flux deviation band from the PDC-3 Axial Offset Control Methodology or to 78 percent rated thermal power when using the 3 percent axial flux deviation band from the PDC-3 Axial Offset Control Methodology. This estimate is still valid for approximately 87 EFPDs of operation as was the previous estimate.
Mr. C. S. Hinnant
- 3 You stated in the letter of October 25, 1996, that a new estimate in PCT will be developed based on (1) the use of the Fuel Cooling Test Facility (FCTF) heat transfer coefficient correlation with the 1986 LBLOCA model modified to eliminate the nonphysical behavior of the 1986 model and (2) limitations of the core peaking factors rather than the core power. This new estimate was reported in your letter of October 29, 1996. The new heat transfer correlation is a linear interpolation of the heat transfer coefficients within the reflood rates between 1 in/sec and 1.77 in/sec. Based on the compensatory measures of the allowable heat flux hot channel factor of 2.40 and nuclear enthalpy rise hot channel factor of 1.73 reported in the October 14, 1996, letter, the new estimated PCT is 2163 oF.
Further, you stated that the new estimate justifies 100 percent rated thermal power. You also explained that the plan mentioned in the October 25, 1996, letter, to develop a new estimate of PCT based on the conservatism due to improvements in fuel pellet manufacturing has been cancelled because the new linear heat transfer coefficient correlation has justified plant operation at full power. This estimated PCT, however, is still valid only for approximately 87 EFPDs of operation, as was the previous estimate.
We have reviewed your letters of October 14, 25, and 29, 1996. We conclude that the modified 1986 model using the heat transfer coefficient correlation discussed in the letter of October 29, 1996, by assuming a linear interpolation of the heat transfer coefficient in the range of reflood rates between 1 in/sec and 1.77 in/sec, is conservative with respect to the measured values and corrects the nonphysical behavior of the 1986 model that we identified in our letter of October 11, 1996. Therefore, we conclude that this model is an acceptable model, in accordance with 50.46(a)(1)(i), for determining that HBR meets the acceptance criteria of 50.46.
Although the staff has accepted your justification for continued operation of HBR up to 87 EFPDs with use of the 1986 model (i.e., a linear interpolation of the heat transfer coefficients within the reflood rates between 1 in/sec and 1.77 in/sec), the NRC staff has determined that it is necessary to review in more detail how the reflood heat transfer coefficient correlation is applied using the SPC methodology for the LBLOCA. Accordingly, we request that you submit the information described in the enclosed request for additional information (RAI) to permit the NRC staff to determine if the application of the 1986 model for HBR is consistent with the database on which the reflood heat transfer coefficient correlation is based. We also request that you address how you plan for HBR to comply with 50.46 for power operation after 87 EFPDs.
Mr. C. S. Hinnant
- 4 You are requested to provide the response to this RAI within 45 days of the receipt of this letter.
Sincerely, (Original Signed By)
Brenda Mozafari, Project Manager Project Directorate II-1 Division of Reactor Projects -
I/II Office of Nuclear Reactor Regulation Docket No. 50-261
Enclosure:
Request for Additional Information cc w/encl: See next page Distribution Docket File PUBLIC PDII-1 RF S. Varga J. Zwolinski E. Merschoff OGC ACRS DOCUMENT NAME: G:\\Robinson\\ROB96355.LTR OFFICE
- PDII-1 LA:PDI, L
BC D:PDII-1 NAME ozafari Dunnin Dr JDon w
iW W MReinhart, DATE 12//} /96
- 12) \\t/96 11/,/96 1-A 12/
/9 COPY YesjNo (Yes No No
__eNe/No OFFICIAI-ECORD COPY-
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE 10 CFR 50.46 LARGE BREAK LOSS OF COOLANT EVALUATION MODEL CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 We request that the licensee provide the following information for the large break loss-of-coolant accident:
- 1. Maximum linear heat generation rate
- 2. Reflood rate as a function of time
- 3. Core collapsed level as a function of time
- 4. Quench time as a function of core height
- 5. Core pressure as a function of time
- 6. Core subcooling as a function of time
- 7. Heat transfer coefficient for peak cladding temperature (PCT) location as a function of time
- 8. Cladding temperature for PCT location as a function of time Enclosure
Mr. C. S. Hinnant H. B. Robinson Steam Electric Carolina Power & Light Company Plant, Unit No. 2 cc:
Mr. William D. Johnson Mr. Dayne H. Brown, Director Vice President and Senior Counsel Department of Environmental, Carolina Power & Light Company Health and Natural Resources Post Office Box 1551 Division of Radiation Protection Raleigh, North Carolina 27602 Post Office Box 27687 Raleigh, North Carolina 27611-7687 Ms. Karen E. Long Assistant Attorney General Mr. Robert P. Gruber State of North Carolina Executive Director Post Office Box 629 Public Staff - NCUC Raleigh, North Carolina 27602 Post Office Box 29520 Raleigh, North Carolina 27626-0520 U.S. Nuclear Regulatory Commission Resident Inspector's Office Mr. Max Batavia, Chief H. B. Robinson Steam Electric Plant South Carolina Department of Health 2112 Old Camden Road Bureau of Radiological Health Hartsville, South Carolina 29550 and Environmental Control 2600 Bull Street Regional Administrator, Region II Columbia, South Carolina 29201 U.S. Nuclear Regulatory Commission 101 Marietta St., N.W., Ste. 2900 Mr. J. Cowan Atlanta, Georgia 30323 Vice President Nuclear Services and Environmental Mr. Dale E. Young Support Department Plant General Manager Carolina Power & Light Company Carolina Power & Light Company Post Office Box 1551 - Mail OHS7 H. B. Robinson Steam Electric Plant Raleigh, North Carolina 27602 3581 West Entrance Road Hartsville, South Carolina 29550 Mr. Milton Shymlock U. S. Nuclear Regulatory Commission Public Service Commission 101 Marietta Street, N.W. Suite 2900 State of South Carolina Atlanta, Ga. 3023-0199 Post Office Drawer 11649 Columbia, South Carolina 29211 Mr. R. M. Krich Manager - Regulatory Affairs Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Unit No. 2 3581 West Entrance Road Hartsville, South Carolina 29550